Purification of contaminated liquid

ABSTRACT

A liquid containing radioactive ions is purified (decontaminated) by contacting the same with an inorganic ion exchange composition having ion exchange sites which can be occupied by the radioactive ions from the liquid. The ion exchange composition is a mixture of an ion exchange medium and an additive which is relatively inert to the ion exchange process and which is a sintering aid for the ion exchange medium designed to lower the sintering temperature of the ion exchange composition. The ion exchange composition may be disposed within a suitable container (e.g., cannister), e.g., made of 304L stainless steel or Inconel 601 and the ion exchange process may be carried out in such container. Alternatively, the ion exchange medium can be employed without being previously admixed with the additive. The additive, if desired, can be admixed at a later stage with the contaminated medium. Thereafter, the mixture may be sintered and disposed of in any desirable manner as by underground burial of the spent mixture within the container. Also, the container may be placed within a suitably designed furnace for carrying out the ion exchange process, sintering of the ion exchange composition and its safe disposal. Methods are also described for making a homogeneous mixture of the ion exchange medium and the additive which, for example, have a certain defined density and particle size relationship.

This is a divisional of application Ser. No. 444,176, filed Nov. 24,1982, now U.S. Pat. No. 4,591,455.

FIELD OF INVENTION

This invention relates generally to the removal of radioactive ions fromliquids containing the same.

In one aspect, the invention relates to compositions having ion exchangecapacity useful, for example, in the removal of radioactive ions fromliquids containing the same.

In another aspect, the invention relates to methods of making suchcompositions.

In still another aspect, this invention is concerned with a method ofremoval of radioactive ions from a liquid containing same using therelatively low melting mixtures comprising ion-exchange inorganicion-exchange media.

In yet another aspect, this invention relates to a novel and uniqueapparatus designed for carrying out the decontamination ion exchangeprocess described herein, which process includes the step of sinteringthe resulting contaminated low melting inorganic ion exchange mediawithin the apparatus.

Other aspects of the invention will become apparent from the ensuingdetailed description of the invention.

BACKGROUND OF THE INVENTION

Disposal of radioactive wastes has become an increasingly difficult taskpresenting serious impediments to the development and utilization ofnuclear power facilities as an alternate source of energy. Largequantities of toxic materials such as high level radioactive wastes areoften stored in spent reactor storage pools, or generated duringreprocessing of spent power reactor fuel or in the operation andmaintenance of nuclear power plants. These radioactive wastes must bedisposed of safely and efficiently.

The difficulty in the disposal of radioactive waste is even more acutewhen the concentration of the radioactive moieties exceed about onemicrocurie per cubic centimeter of waste stream and the radioactivespecies have a multiyear half-life necessitating immobilization of theradioactive moities for a period greater than a century.

One of the most generally accepted procedures for the disposal of suchwastes is to convert the radioactive waste to dry solid form so that thewaste is rendered chemically, thermally and radiolytically stable.

Organic ion exchange media have been employed for the removal of theradioactive ions contained in such wastes but, due to their lowtheshhold for radiation damage, they are not suitable for this purpose.The ultimate radioactive loading level of organic ion exchange media ismore limited than inorganic ion exchange media since the former issusceptible to radiation damage at much lower dosage than the inorganicion exchange media.

Among the inorganic ion exchange media which have been used in theremoval of radioactive ions from radioactive wastes, those based on aporous glass matrix as described in copending application Ser. No.039,595, filed May 16, 1979 and copending application Ser. No. 065,572filed Aug. 10, 1979, have proven to be most useful. Zeolites and sodiumtitanates have also achieved some degree of acceptability as inorganicion exchange media for this purpose.

In general, the method employed for the removal of the radioactive ionsfrom liquids (e.g., radioactive waste stream) containing the samecomprises passing the waste stream through a suitable container(preferably made of stainless steel) containing the ion exchange mediumuntil the ion exchange capacity of the medium is essentially used. Oncethe ion exchange capacity of the medium has been used, the problembecomes one of safe disposal of the container containing the spentmedium without attendant radiation damage and hazards.

Heretofore, one such disposal procedure for the so-called "low level"radioactive waste streams involved draining the container, sealing andburying the same underground in shallow sites (usually less than aboutsix meters deep).

If the container is improperly sealed, there is the likelihood, anddanger, that during trnasportation to the burial site, the seal maybreak, causing radioactive material to be scattered into the surroundingareas. In addition, and in areas having extensive rainfall, the steelcontainer will, with passage of a relatively short period of time, rustand thus expose the spent mass to water which leaches the radioactiveions, especially Cs¹³⁷, Cs¹³⁴, Sr⁹⁰, Co⁶⁰, etc. Consequently, the groundwater will become contaminated with these ions.

Another method of disposing of the spent ion exchange media involvestheir solidification with bitumen. However, this procedure causesradiation damage to the bitumen and is a fire hazard.

The spent ion exchange media have also been disposed of by mixing withcement to form concrete and burying the resulting mass. As in the burialmethod heretofore described, however, this procedure also presents theproblem associated with leaching of the radioactive ions andcontamination of the ground water.

The ion exchange media can be placed in a high integrity containerdesigned to last several hundred years. However, during this period, theion exchange media will release gases due to, for example, radiolosisand decay of organic matter. If this gas is not vented, the containerwill explode. If it is vented, ground water will enter the container. Ineither case, it will not perform as a high integrity container.

Another method of disposal of spent radioactive ion exchange medium in aloose, dry-powdered form involves removal of said medium from theion-exchange container and melting it with glass frit to formborosilicate glass. The requirement of handling such loose, dry powderswhich are not only abrasive but also have a significant amount ofrespirable fines, dictates the use of hot cells, complicated out gassingsystems and remote maintenance, all of which translates into anexpensive operation. See, Electric Power Research Institute ReportNumber EPRI NP-1087 SIA 78-414 "Nuclear Waste Management Status andRecent Accomplishments" May, 1979.

Thus, notwithstanding numerous methods which have heretofore beenproposed for radioactive wastes disposal, and a variety of compositionsemployed for decontaminating such wastes, the problem of effectivedecontamination of these streams and safe disposal of the spent mediaresulting from the decontamination process continues to present seriousdifficulties. Consequently, effective utilization of nuclear power plantfacilities still remains dependent on the development of safe andenvironmentally acceptable methods of decontamination of radioactivewaste streams and the disposal of the spent ion exchange mass resultingfrom treatment of these streams.

Accordingly, an object of this invention is to provide a novel method ofremoving radioactive ions from a liquid containing the same whichcomprises contacting said liquid with a novel composition havingion-exchange capability (which removes said radioactive ions from saidliquid), followed by sintering the resulting radioactive compositionbelow the temperature which causes substantial volatilization and escapeof the radioactive species to the atmosphere.

Another object of the invention is to provide a novel method ofdecontaminating a radioactive-containing liquid which comprisescontacting said liquid with an inorganic ion-exchange materialcharacterized by a relatively high sintering temperature and a highcapacity for radioactive species, there-after adding to the resultingcontaminated material an additive that has the effect of lowering thesintering temperature, followed by sintering the resulting admixturewithout disseminating volatilized radioactive species to the atmosphere.

Another object of the invention is to provide a novel compositioncomprising high melting, inorganic ion-exchange material admixed with asufficient amount of an additive, the resulting admixture having asintering temperature significantly lower than the ion-exchange materialper se.

It is another object of this invention to provide a novel method ofmaking the aforesaid inorganic ion-exchange admixtures, in particular,homogeneous admixtures.

Another object of the invention is directed to novel articles ofmanufacture, in particular, novel containers capable of withstandinghigh sintering temperatures and which include with their structure theaforesaid novel inorganic ion-exchange admixtures.

It is another object of the invention to provide novel compositionscomprising radioactive inorganic ion-exchange material plus an additivehereinafter defined, said compositions characterized by relatively lowsintering temperatures whereby substantial dissemination of theradioactive species in said material is prevented during sintering ofthe compositions.

It is also an object of this invention to provide a novel apparatus forremoving radioactive species from a nuclear waste stream, said apparatusincluding within its structure the aforesaid novel compositions whichcan be sintered within the environment of said waste stream.

SUMMARY OF THE INVENTION

In accordance with one aspect of the present invention a liquid, inparticular, a liquid waste containing radioactive species aredecontaminated by the removal of said ions therefrom. Thedecontamination process employs an inorganic ion exchange material whichhas ion exchange affinity for the radioactive species in the saidliquid.

The process basically involves passing the radioactive liquid, e.g.,nuclear waste stream, through a container of inorganic ion-exchangematerial, e.g., a column or a cannister, whereby the radioactive speciesare trapped within the material. Thereafter, the radioactive-containingion exchange materials can be dried, e.g., by vacuum, heating, etc.,followed by sintering said material to immobilize, encapsulate and/orfix the radioactive species within the sintered body without causingsignificant dissemination of such species to the atmosphere.

The ion exchange composition employed in one aspect of thedecontamination process can be a mixture of an inorganic ion exchangemedium and an additive (further characterized hereinafter) which isrelatively inert in the decontamination step of the novel process.

By the practice of the invention(s), there results several advantagesviewed from costs, environment and energy standpoints. The admixture ofcontaminated ion-exchange medium and additive is characterized by arelatively low sintering temperature as compared to the contaminatedion-exchange medium. At such relatively low temperature there is asubstantial reduction in the sublimation of radioactive species into theenvironment and a concomittant savings in energy. Also, there aresavings in the fabrication of the container when, e.g., the use ofstainless steel, such as 304 L, in lieu of the more expensive Inconeland the like metals is used.

Broadly, the ion-exchange material is characterized by non-radioactive,replaceable ions capable of being exchanged by radioactive ions from theliquid.

Porous silicate glass or porous silica gel, natural or synthetic clays,hydrated metal oxides and alkali metal salts of such hydrated metaloxides, such as sodium titanate, may constitute the ion exchange medium.

BRIEF DESCRIPTION OF DRAWINGS

FIG. 1 is a vertical, cross-sectional view of a container in accordancewith the present invention.

FIG. 2 is a horizontal, cross-sectional view of the container of FIG. 1taken along the line 2--2.

FIG. 3 is a graph of the fraction of original CS¹³⁷ activity presentover time for portland cement, pozzolanic cement and an ion exchangecomposition of the present invention.

FIG. 4 is a schematic representation of an apparatus for practicing theinvention.

DETAILED DESCRIPTION OF THE INVENTION

This invention relates to: (1) an inorganic ion exchange media useful inremoving radioactive moieties from water; (2) an additive which allowsthe ion exchange media to be sintered to a monolith at a relatively lowtemperature thus entrapping the radioactive and/or toxic moieties in ahighly durable waste form suitable for disposal; (3) an ion exchangecontainer suitable for (a) ion exchange, (b) sintering of theadditive+ion exchange media and (c) disposal of the waste form includingmethods of carrying out the sintering in those cases where the ionexchange media has become highly radioactive due to ion exchange; (4)methods of producing the intimate mixing of additive and ion exchangemedia necessary to insure adequate sintering; (5) methods of disposal ofthe sintered mixture and container.

The apparatus of the present invention is comprised of an ion exchangemedia, through which a liquid may be passed for the removal ofradioactive ions, and which is designed so as to permit heating of theion exchange media to form a durable material in which the removedradioactive ions are fixed, and further provides that such apparatus maybe operated within a liquid pool, which may serve as a means ofradioactive shielding and a means to cool the outer surfaces of theapparatus.

The apparatus is basically comprised of an ion exchange media, acontainer for the media called the inner container, a heating means, andin some preferred configurations, an outer container and/or a liquidshielding pool. The ion exchange media is described below.

The purpose of the inner container is to hold the ion exchange media andto provide a physical barrier between the radioactive wastes and theenvironment. The container preferably will be able to withstand possibledamage during an accident in shipping or storage as well as be able toendure the heating to sintering of the ion exchange media.

The choice of the canister materials is dictated primarily by therequirements of high temperature strength, resistance to corrosion fromthe molten ion exchange media, resistance to stress corrosion cracking,ductility, and toughness. Stainless steels, such as 304L Stainless Steelhave been examined and found to be useful for this purpose. Also, anumber of corrosion-resistant alloys of Ni and Cr such as Inconel 601have also been found to be suitable. Other similar alloys also would besuitable.

304L Stainless Steel has the advantages of lower cost, ease offabrication, machining and welding, but has a lower upper usetemperature and less resistance to stress corrosion cracking thanInconel 601.

The strength of 304L Stainless Steel is about 60% of that of Inconel 601up to about 1050° C. which is a practical upper use temperature for the304L Stainless. Above this temperature, excessive oxidation of 304LStainless occurs in air, and in vacuum its strength is loweredapparently because of the absence of nitrogen. Inconel 601 does notsuffer the corrosion problems and its use can be extended up to about1150° C. With these temperature limits and published creep strengthdata, one practiced in the art can design a cannister which willwithstand the requirements for processing of radioactive wastes asdescribed herein. Further, such a cannister can meet the requirementsfor shipping and extended storage of the solidified waste.

The choice of alloy to be used is dictated by the sintering temperatureof the ion exchange media-additive mixture (described below), with 304LStainless Steel being preferred for temperatures below about 1050° C.,because of its lower cost, and Inconel 601 being preferred above thistemperature because of high temperature strength and corrosionresistance.

In a separate embodiment of the invention, the heating means may be anintegral part of the apparatus or may be an independent apparatus. Theheating means may be a resistance heater or an RF (radio frequency)heater. If the heating means is an integral part of the apparatus,resistance heaters are preferred for cost and versatility since they canprovide, at lowest cost, temperatures to 1200° C. and can be configuredto provide zone heating if desired. Resistance heaters are discussedbelow. If the heating means provides an independent heat source, anyconvenient heat source, e.g., resistance heaters, may be used, but RFheating may be preferred because of its low maintenance cost and itsversatility. In such a case, the container, which is a metal, can serveas a susceptor for the RF power.

The optional outer container serves as a further barrier, in addition tothe inner container, between the radioactive wastes and the environment.for this reason, its design and the choice of container materials isgoverned by similar constraints as for the inner container. If theapparatus is to be operated in a liquid pool, and has an integralheating means, the outer container acts as a barrer between the pool andthe heating means. Further, the outer container may be required towithstand the pressure exerted by the liquid when the apparatus issubmerged and the interior of the apparatus is evacuated and the heatingmeans is activated.

To illustrate further details of the apparatus, one preferred embodimentof the invention is shown in FIGS. 1 and 2. FIG. 1 illustrates in acutaway view (with the cut being a plane passing through the axis of thebasically cylindrical apparatus). The identifying numbers are the samein the two figures. FIGS. 1 and 2 illustrates major components of theapparatus to which, in operation, other components and design detailswould be added by those pr acticed in the arts of mechanical design,heat transfer, and temperature control.

The apparatus is comprised of an outer jacket 10 which may be of anyconvenient shape, which is preferably a hollow right circular cylinderwith closed ends. The outer jacket is preferably made of metal such as acarbon steel, stainless steel, Inconel or other suitable metals capableof sustaining the temperatures reached during the heating of the ionexchange material and having sufficient mechanical strength to withstandthe pressure exerted by the liquid when the apparatus is evacuated. Thethickness of the metal and the details of the shape can easily bedetermined by one practiced in mechanical design. In one embodiment ofthe invention, the metal is 304L stainless, preferred because of itsrelatively low cost and high durability. In one embodiment of theinvention the inner surfaces of the outer jacket 10 are polished, or arecoated with a reflecting coating or lined with a reflective film toreflect thermal radiation. The wall and ends of the jacket may betraversed by pipes, electrical feedthroughs or other components, some ofwhich are illustrated in FIG. 1 and are discussed below.

The heating assembly 11 is contained within the outer jacket 10. Betweenthe walls of outer jacket 10 and the heating assembly 11 is the space12, which can be evacuated to reduce heat transfer from the heatingassembly 11, to the outer jacket 10 or it may be filled or partiallyfilled with supports or insulating materials, or it may be filled withgas during part of the operation of the apparatus to promote heattransfer from the central region of the apparatus to the outer jacket10. In the preferred embodiment, one or more heater support spacers 13,made of insulating refractory material, are in the space, 12. Inaddition pipes, other supports, electrical leads or other apparatus maybe in or pass through the space, 12, as discussed below and as requiredby those practiced in design.

The heating assembly, 11, may be comprised of any suitable electricheater or plurality of heaters with resistance heaters being preferred.In a preferred embodiment, the heating assembly is basically cylindricalin shape with its axis essentially coicident with that of the axis ofouter jacket 10. The outside diameter of the assembly 11 is less thanthat of the outer jacket 10, and the inside diameter of 11 is greaterthan that of the inner jacket 14. The length of the assembly is lessthan that of the length of outer jacket 10, and preferably greater thanthe length of inner jacket 14. In a preferred embodiment, the heaterassembly is comprised of more than one heater element. Each element ofthe assembly in this embodiment is a right circular cylinder with adiameter the same as the whole assembly. A preferred element is aKANTHOL type A1 heater wire embedded in a cylindrical ceramic substrate.In the preferred embodiment, each element has independent electricalpower leads 22 and a separate temperature sensor or sensors which arepreferably thermocouple probes 21. In the preferred embodiment, separatetemperature control of elements permits the ion exchange media 15 to beheated in a zone manner. The zone is defined by the volume within aparticular heater element. In the preferred embodiment, during theheating stage of the process, the zones are heated in such a mannerthan, if desired, the ion exchange media may be heated starting from thebottom zone and progressing upward to the top zone in a programmedmanner as determined by the drying and sintering characteristics of theion exchange media 15. The temperature controls for the heater assemblyare located remotely from the apparatus.

The power leads 22 and thermocouple probes 21 pass through the space 12and pass through the wall of outer jacket 10 via suitable electricalfeed throughs 24, compression feed through 25 or other suitable feedthrough means as selected by those practiced in the art. The innerjacket 14 is a container of any convenient shape and of suitablematerials as discussed previously with 304L stainless steel beingpreferred, and in one preferred embodiment, is essentially a rightcircular cylinder with closed ends, with its axis essentially coincidentwith the axis of 11, the heater assembly. The inner jacket 14 issupported by a plurality of supports 51. The inner jacket 14 ispenetrated by at least two pipes 26 and 27, which are the inlet andoutlet pipes, respectively. At the traversal point, the pipes are sealedin a lead-tight fashion, preferably by welding. The inlet pipes passesthrough the sintered disc filter 30 which is a stainless steel sintereddisc. The disc is weld sintered to the pipe at the traversal point. Theinlet pipe also extends through the media 15 and terminates above thetop end of the ion exchange media. In one embodiment of the invention,the ion exchange media is constrained at the top by a second stainlesssteel sintered disc 32 through which the inlet tube 26 also passes andto which it is sealed at the traversal point. The sintered discs 30 and32 are welded to the inner surface of the inner jacket 14.

In one embodiment of the invention, the outlet tube terminates above theupper sintered disc 32. In a preferred embodiment, the outlet pipe 27 isessentially an inverted U shape and traverses the lower sintered disc 30where it is weld sealed at the traversal point, then passes through theion exchange media 15, passes above the top of the ion exchange mediaand through the upper sintered disc 32, forms a U and returns throughthe upper sintered disc 32, through the ion exchange media 15, the lowersintered disc 30 and terminates in the space between the loweredsintered disc 30 at the bottom of the inner jacket 14. The pipe 27 isweld sealed to the sintered discs 30 and 32 at each traversal point.

Outside the outer jacket 10, the pipes 26 and 27 are terminated bysuitable closures such as valves 40 and 41, which in one preferredembodiment are air operated valves.

Within the inner jacket 14, there may be supports for the disc 30 suchas the supports 50 which are a plurality of posts welded to the bottomof the inner jacket 14, and/or to the disc 30. In addition, the innerjacket 14 may be supported with respect to the outer jacket 10 bysupports 51.

The outer jacket 10 is in one preferred embodiment, traversed by pipe 28which serves as a vacuum or gas port, which is sealed at the traversalpoint to the outer jacket 10 and which terminates outside outer jacket10 by a valve or suitable closure 42. The outer jacket 10, in apreferred embodiment, may optionally be traversed by a vacuum feedthrough such as thermocoupled gauge tube 60 which is sealed at thetraversal point by a suitable means such as pipe threads, welding orcompression fitting.

FIG. 2 shows a cross-section view of the apparatus showing the heattransfer fins 70 which are preferably made of metal and which are notshown in FIG. 1 for clarity. The fins fit in the space between thesintered discs 30 and 32 or are above the lower disc if only one disc ispresent. The heat transfer fin assembly is held together at the top andbottom by rings 72 which are welded to the fins at the points ofintersection. In addition to the mechanical supports and spacersindiated, additional such supports and spacers may be added by onepracticed in the art to provide mechanical stability and strength to theapparatus. Further, the shapes and construction of the inlet and ouletpipes 26 and 27 and the outer jacket 10 and inner jacket 14 mayincorporate design features to limit stress during the heating steps. Tofurther explain the function of the components of the apparatus, thetypical operation of the preferred apparatus is described.

Before operation, the ion exchange media is properly prepared as isdescribed below. The various utilities, sensor controls, inlet andoutlet pipes are connected. Valves 40 and 41 are opened and theradioactive liquid flows through inlet pipe 26, passes through the uppersintered disc 32, passes through the ion exchange media 15 and throughthe lower sintered disc 30 an is forced out through the outlet tube 27.The apparatus may be submerged in a water pool for shielding. This poolcan be the fuel element storage pool of a reactor. The purpose of theinverted U shape of 27 is to prevent molten radioactive ion exchangemedia, produced during the heating stage, from reaching the outlet valve40 in the event of accidental leaking of the sintered disc during theheating stage.

The radioactive liquid flows through the ion exchange media 15 whereuponthe radioactive ions are removed. the effluent from the apparatus ismonitored to verify the removal of the radioactive ions and to indicatewhen the capacity of the ion exchange media 15 for removal of the ionshas been reaches. If required, the effluent may flow through a secondapparatus or filter for further polishing of the liquid before theliquid is released for disposal or further use.

When all the liquid has been cleaned, or the capacity of the ionexchange media 15 has been exhausted, the flow of radioactive liquid isstopped. Clean water is then flowed through the apparatus entering 26and exiting 27 to wash the pipes and valves 40 and 41. The inner jacket14 is then evacuated to remove excess liquid in the ion exchange media.The outer jacket 1 may be evacuated through a vacuum port 28 at thistime or later during the heating steps. Sufficient ambient temperaturevacuum drying may be accelerated by heating the ion exchange media viathe heater assembly 11. In the preferred embodiment, the temperature ofthe ion exchange media is raised by zones, with the lower portion of theion exchange media being heated first. By heating to sintering from thebottom and proceeding toward the top, volatilization of radioactivematerials can be minimized. Material volatilized in the hot lowerregions will recondense in the cooler upper regions with the result thatonly a small percentage of all the volatilized material will escapebefore the top region of the media is finally sintered.

After sintering, the inner jacket 14 may be rapidly cooled by allowinggas to enter the space 12 via the vacuum port 28. After the apparatus iscooled, the utilities, sensors and pipes may be disconnected and theapparatus may be prepared for storage or transportation to a storagesite.

According to the present invention, an inorganic ion exchange media isused to remove radioactive ions from solution. The ion exchange mediaused in the invention is a porous inorganic structure having a surfacearea (as usually measured by BET) greater than about 20m² per gram orcm³. In the case of cation exchangers the media will exchange H⁺, alkalimetal ion, alkali earth metal ions, group IB or IIB metal ions orammonium ions for the radioactive ions in solution. Since it is possibleto tie organic molecules to a porous silica glass matrix, the media mayhave chelating functions in its surface (U.S. Pat. No. 4,333,847). Foranions, the media usually has available OH⁻ or Cl⁻ for exchange;however, F⁻, NO₃ ⁻, and other anions may be used. Also, it can havesites which will complex with specific atoms in solution whether theatoms are in the form of cations, or anions or are neutral.

The inorganic ion exchange media may generally be broken down in thefollowing classes: (i) silica based materials including glasses, silicagels and other synthetic and natural silicas; (ii) clays and zeolites,including both natural and synthetic varieties, (iii) hydrated metaloxides including phosphates, molybdates, vanadates, and others; and (iv)alkali metal salts of the above.

Silica gels have long been used as ion exchange media in a number ofapplications such as chromatographic packing (U.S. Pat. Nos. 3,722,181and 3,983,299), removal and separation of fission products (U.S. Pat.Nos. 2,717,696 and 2,855,269), and general ion exchange (Patrick andBarclay, Ponomareva et al). Silica glass based media has been used forgeneral removal of a wide range of cations (Ser. No. 370,437, filed Apr.21, 1982) and anions (U.S. Pat. No. 4,333,897) found in radioactivewaste and for removal of organic pollutants (U.S. Pat. No. 3,901,818).Silica gels also have been found useful in chromatographic packing (U.S.Pat. Nos. 3,722,181 and 3,983,299). Various other silicates have beenproposed for radionuclide waste entrapment (U.S. Pat. Nos. 3,959,172,3,451,940 and 3,849,330).

Clays were perhaps the earliest inorganic ion exchange media and havebeen applied in numerous processes, among them disposal of radioactivewaste (U.S. Pat. Nos. 3,274,784, 3,925,992 and 3,093,593). Examples ofclays that have been used as ion exchange media are bentonite,muscovite, vermiculite, kaonite, illite, montmorillonites, andnontronite. Zeolites are generally porous sodium alumino silicates thatare either natural (e.g., chabazite, clinoptilolite, mordenite,erionite) or synthetic (e.g., those produced by Union Carbide Corp.).Some titano and zircano silicates (U.S. Pat. No. 3,329,481) are alsoclassified as zeolites. These have been used as molecular sieves forpurifying gases, for radioactive waste purification (U.S. Pat. No.3,167,504) and for ammonia removal from waste water.

A number of hydrated metal oxides have been shown to be effective media.Among these are alumina, titania, zirconia (U.S. Pat. No. 3,101,998),thoria and various phosphates (e.g., zirconium phosphate, tantulumphosphate), vanadates and other complex oxides. In some cases theseoxides are combined with silica or other binders to provide particlestability. Hydrated metal oxides have been used for removal ofradioactive ions (U.S. Pat. Nos. 3,337,737, 3,338,034, 3,522,187 and2,859,093).

The sodium salts of the hydrated metal oxides also can be used asinorganic ion exchange media. Sodium titanate has been shown to beparticularly effective for the removal of radioactive strontium fromsolution.

In the present invention, the function of the additive is to act as asintering aid, which lowers the temperature at which a suitable granularpowder will sinter. By sintering, we mean heating to a temperaturewhereby the individual grains of the powder stick together and flow intoeach other. As sintering progresses the volume of the granular powderwill contract as porosity is reduced. A well sintered material may stillhave some porosity, but it will be discrete rather than interconnected.A well sintered material will be referred to herein as a monolith todistinguish it from the original powders. Since the expansioncoefficient of a container in which the sintering is carried out may bedifferent from that of the sintered material, the sintered material mayfracture upon cooling. Thus, monolith refers to unbroken or brokenpieces of ceramics, glassy, and/or partially or totally crystallizedmaterial, made up of the original ion exchange medium and additive whichhas been heated until sufficient volume reduction has occurred todisrupt the interconnected porosity.

Inorganic ion exchange media, such as those mentioned above, can besintered without a sintering aid. However, the ordinary processtemperatures required for sintering are high enough to vaporize anddrive off radioactive substances that may be trapped on the ion exchangemedia, creating a major pollution hazard. The temperatures required tosinter the ion exchange media are also high enough to destroy mostreadily available materials that could be used for the ion exchangecolumn cannister. The additive lowers the sintering process temperatureby performing one or both of the following functions: (i) In the casewhere the ion exchange media forms glass grains (either because it wasoriginally glassy or because upon heating to the process temperature, ittransformed itself into an amorphous phase) the additive lowers theviscosity and permits the grains to flow into each other. (ii) In thecase where the ion exchange media has crystalline grains, the additivecan form an eutectic mixture which will lower the liquidous temperaturebelow the desired process temperature, thus causing the whole mess,additive plus ion exchange media, to melt. Note that without thepresence of the additive the ion exchange media would have a liquidoustemperature higher than the desired process temperature. If aninsufficient amount of additive is present to lower the liquidous of thewhole mass, a thin liquid layer between the crystal grains can still beformed. Under such conditions mass transfer between regions where thesolubility of the crystals is slightly higher, to those regions where itis slightly lower, can still lead to sintering. In either process, (i)or (ii), the sintering time is determined by the viscosity of the fluidadditive plus dissolved ion exchange media; the lower the viscosity, theshorter the sintering time. Thus, it is important that the additivelower the viscosity.

The ideal additive has the following properties: (a) It lowers thesintering temperature of the ion exchange media+additive below themaximum allowed process temperature, which is determined by the lowestof the following two temperatures: (1) the maximum temperature the ionexchange column cannister can withstand or (2) the temperature at whichthe radioactive substances trapped on the ion exchange media volatizeappreciably. (b) It lowers the viscosity of the ion exchangemedia+additive mixture sufficiently to give good sintering in areasonable amount of time. (c) It does not deleteriously affect the ionexchange capacity and efficiency of the ion exchange media. (d) Iteasily mixes with the media to form a uniform, homogeneous mixture. (e)It sinters with the ion exchange media to form a highly durablemonolith. These properties will be discussed in more detail below.

As discussed above, currently used cannister materials can ordinarilywithstand maximum preferred process temperatures of less than about1200° C. The temperature at which the radioactive substances trapped onthe ion exchange media volatilize appreciably, of course, depends on thenature of those substances. For example, radioactive cesium volatilizesappreciably at temperatures above 1050° C. Radioactive substances suchas uranium, plutonium, strontium and cobalt would not be expected tovolatilize until temperatures considerably higher than 1050° C. arereached, whereas a substance such as iodine would be expected tovolatilize at temperature well below 1050° C. On the basis of the abovediscussion, it is noted that the maximum allowable process temperaturefor sintering will be in the neighborhood of 1050° C. or lower. Theexamples provided below demonstrate that a variety of additives can beused to produce good sintering for a variety of inorganic ion exchangemedia at temperatures of 1050° C. and below. Without additives, the ionexchange media of the invention would sinter at temperatures in excessof 1300° C.

As noted above, the required sintering time is determined by theviscosity of the fluid additive plus dissolved ion exchange media, thelower the viscosity the shorter the sintering time. In general sinteringoccurs in a matter of hours when the viscosity is below 10⁶ Poise. Formost substances viscosity can be lowered by raising the temperature.

For example, the viscosity of the glass used as Additive B, illustratedin the examples, is halved when the temperature is increased from 1000°C. to 1050° C., whereas the viscosity of Additive C drops by almostthree orders of magnitude as the temperature rises from 720° C. to 1050°C.

The most important viscosity as it relates to the present invention isthat of the additive+ion exchange media mixture. We have discoveredthat, for a given ion exchange media, good sintering at 1050° C. can beproduced by using an additive having a viscosity below about 10³ Poiseat 1050° C., preferably below 10² Poise. The Examples illustrate thatthe additives that meet the requirement give good sintering at anadditive volume of 20 percent. The additives which fail to meet thisrequirement fail to produce good sintering.

Three other factors controlling the sintering can also be illustratedusing the Table below, which summarizes the examples.

                  TABLE I                                                         ______________________________________                                                        Viscosity of       Sin-                                               Addi-   Additive at        tering                                                                              Perfor-                              Addi-   tive    Sintering  Sintering                                                                             Time  mance*                               tive    Vol %   Temp (Poise)                                                                             Temp (°C.)                                                                     (min) Index                                ______________________________________                                        1   A       50      10.sup.2 1050    180   3                                  2   A       20      10.sup.2 1050    300   3                                  3   B       50      10.sup.5 1100    15    1                                  4   B       50      10.sup.4 1250    15    3                                  5   B       20      10.sup.5 1100    15    0                                  6   C       20      10.sup.3 1050    180   2                                  7   D       20               1050    180   2                                  8   E       20      1        1000    60    3                                  9   F       33      10.sup.4 1050    10    0                                  ______________________________________                                         *3 = good sintering                                                           2 = medium sintering                                                          1 = poor sintering                                                            0 = no sintering                                                         

The first factor to be considered is that raising the temperature lowersthe viscosity of the mixture and promotes sintering. This effect can bedemonstrated for the mixture by comparing the results for 50 volume %Additive B at 1100° C. and 1250° C. (lines 3 and 4 of Table I). Thelower temperature gives poor sintering whereas the higher gives goodsintering. The second factor is that sintering improves as thepercentage of additive increases. Comparing lines 5 and 3 of Table I, itcan be seen that increasing the volume percent of B from 20 to 50improves the sintering performance from no sintering to poor sintering.The third factor controlling the sintering process is that the sinteringtime can be reduced by increasing the percentage of additive. This isillustrated by lines 1 and 2 of Table I. The sintering time drops by 3/5as the volume of Additive A increases from 20 to 50%. Increasedpercentages of additive thus lower the mixture viscosity. This improvessintering allowing shorter sintering times or lower sinteringtemperatures. Unfortunately, as discussed below, increased percentagesof additive have a deleterious effect on ion exchange capacity if theadditive is present during the ion exchange process.

Additives that give good sintering properties to the mixture generallyhave compositions that include high percentages of all or some of thefollowing coponents: alkali metal oxides; alkali earth oxides; B₂ O₂ ;P₂ O₅ ; PbO; B₂ O₃. Additives that have high percentages of SiO₂, Al₂ O₃or other high melting refractories such as ZrO₂ generally impart poorsintering qualities to the mixture.

If present during the ion exchange process the additive can affect theion exchange capacity and efficiency of the mixture in a variety ofways. The additive acts to dilute the ion exchange media in the ionexchange column. Since the ion exchange capacity of the mixture dependson the amount of ion exchange media present, the higher the percentageof additive, the lower the capacity of the mixture. Thus, if presentduring ion exchange, the volume percentage of additive should be below50%, preferably below 30% and most preferably below 20%.

The additive may also affect the ion exchange capacity of the mixture,if at any time prior to or during the ion exchange process, the additivereleases constitutents which interfere with the removal of theradioactive substances from solution. If, for example, the additivereleases lead ion to the ion exchange media prior to, or during the ionexchange process, the capacity of the ion exchange media to remove leadion from solution could be reduced or even eliminated. The speciesreleased by the additive need not be identical to the species to beremoved from solution for the reduction or elimination of the capacityof the ion exchange media to occur. For example, the release of largequantities of Na⁺ or K⁺ ion would tend to interfere with the removal ofCs⁺ from solution. Similarly the release of Ca⁺⁺ or Mg⁺⁺ would interferewith the removal of Sr⁺⁺ from solution. In both of the above cases, thereleased ions compete for the ion exchanges sites of the ion exchangemedia with the radioactive ions.

The additive may also interfere with the ion exchange media by producinga chemical reaction with species in solution to produce a new species,which the media cannot effectively remove from solution, or byphysically coating the ion exchange sites of the ion exchange mediathereby reducing capacity.

In the present invention, three types of additive are used. The first isa powdered solid which is mixed with the ion exchange media and ispresent during the ion exchange process. This class of additives musthave a very low dissolution rate in order that (a) the additive does notdissolve away during the ion exchange process and (b) the additive doesnot release interfering substance into solution. To illustrate thepermissible dissolution (leach) rate, consider the following situationwhich would be characteristic of an additive powdered to the 355-210 μmsize having a leach rate of 1×10⁻⁵ g cm⁻² d⁻¹, a density in powderedform of 1 g/cm³, and a surface area of 100 cm² /g. Referring to thedissolution rate of the additive during the ion exchange process, thefraction of the additive that would dissolve in one day would be equalto the leach rate times the surface area or 0.001 per day. It would thustake 10 days of passing ion exchange solution over the additive for 1 %of it to dissolve. Thus, dissolution rates should be less than 10⁻⁴ gcm⁻² d⁻¹, preferably less than 10⁻⁵ g cm⁻² d⁻¹, at the operatingtemperature.

Referring to the release of an interfering substance into solution, andassuming the percentage of the additive in the mixture is 20%, then forevery cm³ of mixture there would be 0.2 cm³ or 0.2 g of additive and 0.8cm³ of media. For the leach rate of 10⁻⁵ g cm⁻² d⁻¹, 0.2 mg of theadditive will dissolve in one day per cm³ of mixture. A typicalinorganic ion exchange media has a capacity of about 10 mg of ion/cm³,so that even in the unlikely event that all of the dissolved additiveformed substances that were absorbed on the ion exchange media, only21/2% of its capacity would be lost per day. Thus to prevent deleteriouseffects on the ion exchange media's capacity, the dissolution rate of apowdered solid additive in solution should be less than 10⁻⁵ g cm⁻² d⁻¹,at the operating temperature.

A second type of additive is introduced as a powdered solid after ionexchange has taken place but prior to sintering, thus avoiding theproblem of interference of the additive in the ion exchange process.This type of additive requires that steps be taken to open the ionexchange cannister after ion exchange, to add the additive to thecannister and to mix the additive+ion exchange media to provide auniform, intimate mixture suitable for sintering. This mixing can beobtained mechanically by stirring or by other means described below.

A third type of additive is introduced in solution form. In thisembodiment of the invention the additive is dissolved in a suitablesolvent, which will dissolve appreciable amounts of the additive, andwill evaporate cleanly, leaving the aid behind. When using thisadditive, the ion exchange material is dried, and enough additivesolution is introduced to cover the ion exchange material. The solventis then evaporated leaving behind the additive as a coating on the ionexchange material, or as a fine powder dispersed uniformly throughoutthe ion exchange media. This process may be repeated several times toincrease the amount of sintering aid deposited with the ion exchangemedia.

This type of additive has the advantage that it is not present duringthe ion exchange process. There is thus no dilution of the ion exchangemedia's capacity and no change for chemical or physical interferencewith the ion exchange process.

The liquid additive is an alkali metal oxide, SiO₂, B₂ O₃, PbO, P₂ O₅,Bi₂ O₃, ZnO, CoO, MgO and mixtures thereof. The preferred liquidadditives are B₂ O₃, P₂ O₅, ZnO, CoO, PbO and mixtures thereof. Theadditive is dissolved in a liquid solvent. The liquid solvent is water,lower molecular weight alcohols, organic acids, ammonia and minieralacids. The preferred solvent is water or ammonia.

The ion exchange media and additive must be intimately mixed immediatelyprior to and during the sintering process. In the case where theadditive is dissolved in solution, upon the evaporation of the solvent,the additive will be uniformly dispersed throughout the media. When theadditive is added prior to the ion exchange, process, the intimatemixing of additive and ion exchange media can be obtained by standardmechanical means such as stirring or tumbling. Care must then be takenthat the additive and ion exchange media do not separate during anyprocess prior to sintering. Of crucial importance is the process ofbackwashing which is carried out to remove fines and air pockets fromthe ion exchange media prior to the beginning of the ion exchangeprocess. Backwashing consists of passing water against gravity from thebottom of the column to the top at a rate that is usually sufficient tolift and expand the ion exchange bed. As is illustrated in the Examples,backwashing can lead to separation of the mixture unless precautions aretaken to prevent it. Similar problems can occur during any counter-flow(against gravity) operation.

Separation of additive and ion exchange media, can be prevented in thefollowing ways:

i. The particle size of the two may be adjusted to make up fordifferences in density so that upon backwashing a uniform mixture isformed. An approximate rule of thumb for choosing particle size is givenby:

    ρ.sub.A S.sub.A.sup.1/2 =ρ.sub.M S.sub.M.sup.1/2

where ρ is the bulk density, S the size and A and M stand for additiveand ion exchange media respectively.

ii. The mixture is placed in the cannister and confined in such a waythat it cannot shift or expand during backwash or counter-flowoperation.

iii. Subsequent to the ion exchange process, compressed air is forcedupward through the mixture to produce uniform mixing. However, thismixing method must be very carefully performed to avoid release ofradioactive materials to the atmosphere.

iv. Subsequent to the ion exchange process the ion exchange media andadditive are mixed mechanically by a stirring device such as a paddle.Methods i, iii, and iv are also applicable where the additive isintroduced as a solid powder after the ion exchange process. Methods iand ii are the most preferred embodiments of the invention. Methods iiiand iv are recognized to be standard industrial methods.

After the sintering, the cannister and monolith are disposed of bystorage and or burial. If the cannister is buried on land or at sea, orif during transportation an accident occurs, the monolith willeventually be contacted with water. Such contact will permit thebeginning of leaching. It is the purpose of the present invention tominimize the release to the ecosphere of radioactive elements trapped inthe monolith. Thus, the additive should be selected such that itscomposition improves the chemical durability of the monolith.

The requirement for low viscosity during sintering demands the selectionof components for the additive of alkali metal oxides, B₂ O₃, P₂ O₅,PbO, Bi₂ O₃ or alkali earth oxides, all of which reduce chemicaldurability. The selection of components for improved chemical durabilitySiO₂, Al₂ O₃, and ZrO₂ all produce high viscosities. Thus, thecomposition of the additive has to be a compromise between good chemicaldurability and low viscosity. Glasses containing PbO seem to beespecially suitable in this respect.

The data available on the rates of dissolution of other candidatematerials, in particular cements, proposed for the fixation of low andintermediate-level wastes, are not very extensive. In many cases onlyweight changes have been reported, and these results cannot be appliedto the evaluation of either matrix dissolution rates or waste extractionrates, since weight losses tend to be attenuated and often overshadowedby the strong tendency of the materials to absorb water. In order tomeasure the rate of extraction of components of the cement, it isnecessary to determine the levels of such components in the usedleachant. Data based on extraction tests carried out on typicalconstruction cements show that, in the case of most Portland cements,the extent of lime dissolution is only limited by the concentration ofCa(OH)₂ in saturated aqueous solutions, and that the amount of dissolvedCaO under moderately rapid flow conditions can reach 10% of the totalweight of the cement within a period of a few hours. For Pozzolanic(ash-containing) cements the attack is slower by about one order ofmagnitude than for Portland cements. Recently, data from extractiontests have been developed for radioactive waste fixation. In the case ofimproved Portland Cement, the extraction rate at the end of a 100-daytest was calculated to be 1 mm/year, corresponding to 9.4×10⁻⁴ g cm⁻²d⁻¹. (A density of 2.55 gcm⁻³, characteristic of Pozzolanic cements, isassumed in further calculations.)

The dissolution rates of the monoliths of this invention are generallymuch lower than the cement dissolution rates. For example thedissolution rate of the monolith formed from Additive A+Media Xhereafter called waste form AX is lower than the dissolution rates ofthe two special cements by factors 3000 and 700, respectively. However,in deciding whether various waste-forms are sufficiently effective asbarriers against the dissolution of hazardous components of the solid, acalculation of the amount of dissolution products released into asurrounding medium is much more meaningful than the gross dissolutionrate. Such a calculation in the case of a radioisotope which does nothave a relatively long-lived parent isotope, and of a uniform (i.e., nota multi-barrier) solid waste-form, is based on the equation ##EQU1##where M_(i) is the amount of radioisotope released into the surroundingmedium due to the dissolution of the solid and not yet decayed after acertain storage time t, M_(oi) is the initial concentration of i in thewaste-form, D is the matrix dissolution rate, ρ is the density of thesolid, r_(o) is the radius of the solid waste-form, assuming cylindricalgeometry, and ρ_(i) is the mean lifetime of the isotope, i.e., thereciprocal of the exponential decay rate (not the half-life). Thiscalculation is valid up to the time given by ##EQU2## at which thematrix has completely dissolved; it is possible to obtain M_(i) at anylonger time t>t_(c) from the simple exponential expression

    M.sub.i =M.sub.oi e.sup.-t/τ.sbsp.i

The values of D and for the three waste-forms under discussion here,i.e., AX and the two cements, at a temperature of 24° C., have beengiven below. It is assumed that the loading of the waste-form in thecase of low or intermediate-level waste will not be so high as to causethe surface temperature of the waste to rise considerably. Although thewaste packages will be cylindrical with a typical radius of 30 cm,cracking induced by thermal stresses during the production of the glassand fracturing induced by temperature and humidity fluctuations of thecement will cause the solid waste-form to be broken into pieces ofirregular shape with a typical smallest dimension (equivalent to r_(o)in the cylindrical case) of approximately 1 cm. Each piece will besurrounded by cracks into which water will be able to seep and interactwith the solid.

    ______________________________________                                        Material        D(gcm.sup.-2 d.sup.-1)                                                                    ρ (g cm.sup.-3)                               ______________________________________                                        Portland Cement 9.4 × 10.sup.-4                                                                     3.13                                              Pozzolanic Cement                                                                             2 × 10.sup.-4                                                                       2.55                                              AX              3 × 10.sup.-7                                                                       2.2                                               ______________________________________                                    

The isotope considered here ¹³⁷ Cs, with a half-like of t_(1/2) =30.174years and mean life τ=T_(1/2) /ln2=43.532 years. ¹³⁷ Cs is the majorconstitutent of low and intermediate-level waste streams; ⁹⁰ Sr, with aT_(1/2) of 28.1 years, yields very similar results, while othercomponents, such as ¹⁰⁶ Ru (T_(1/2) =1.0 year) and ⁶⁰ Co (T_(1/2) =5.26years) have much shorter means lifetimes and are usually much lesssignificant. Upon substituting the above value into Eq. (1) (fort≦t_(c)), the three curves A, B, and C, which are shown in FIG. 3, areobtained for improved portland cement, special ash-containing(Pozzolanic) cement, and AX, respectively. It can be seen that in thecase of continuous exposure to water, most of the radioactivityinitially present in the cements will pass into the environment within ashort period (82% after 8.2 years in the case of Portland cement, 52%after 22.6 years in the case of Pozzolanic cement). The maximum amountof extraction for the glass can only reach 0.16% of the initial activityin the solid; this level will be reached after a period of 43.5 years.This period coincides with τ_(i) ; since the time t_(c) required forcomplete dissolution of the glass, 20,200 years, is very long comparedwith τ_(i), whereas the time for maximum extraction is 9 years in thecase of Portland cement. Accordingly, Pozzolanic cement offers noimprovement upon Portland cement at periods longer than 35 years.

Another important conclusion derived from FIG. 3 is that after thewaste-form first becomes exposed to water, the amount of activityreleased to the environment within 6 months (typical time necessary torecover waste after transportation accident) will reach 11% of the totalacivity in the solid in the case of Portland cement, 3% in the case ofPozzolanic cement, and only 0.005% in the case of AX. Furthermore, onceexposure to water has taken place, the activity in the mediumsurrounding the cement will persist at a level exceeding 1% of the totalinitial activity of the solid for the next 200 years. For AX thecorresponding residual activity will only reach 0.02%.

It is important to note that the dissolution rate of AX used in thesecalculations is very conservative. The data, indicate that in the longterm the dissolution rate of AX can be expected to drop by at least afactor of two, under the 3×10⁻⁷ g cm⁻² d⁻¹ level observed after 25 days.It can be concluded that after a relatively short period this glass willcompare favorably with the best types of borosilicate glass proposed forhigh-level waste fixation, for example, Battelle PNL 76-68, which has along-term dissolution rate of 1.4×10⁻⁷ g cm⁻² d⁻¹ at 20° C. (The latterresult was obtained by prolonged immersion of the glass at 70° C. andthen at 45° C. prior to a two-month test at 20° C. which did not showfurther dependence of the dissolution rate on time.)

It is also necessary to consider the possible preferential leaching of¹³⁷ Cs relative to the matrix. In the case of AX containing 0.1% Cs, the29-day immersion test described above showed that Cs levels in the usedleachant always remained below the analytical detection limit of 0.01mg/lit. Results are available from the long-term test on the BattellePNL 76-68 borosilicate glass which show that the normalized leach rateof Cs is only 62% above the silica dissolution rate.

In contrast, the dissolution rates quoted for the cements cannot beconsidered to be conservative and suitable for use as long-term upperlimits. The fixation method based on cements, unlike direct sorptiononto forms such as AX, involves adsorption of the waste constituents inzeolites and mixing the zeolites with cement at a ratio of approximately70:30. Since zeolites are reversible ion-exchange media the chemicaldurability of the combined product and, in particular, the preferentialleach rates of mobil ions, such as Cs⁺, are likely to be much higherthan the dissolution rates of the carefully optimized pure cements.

In addition, the cements have an interconnected open pore structurewhich will allow rapid penetration of ground water into the waste formfollowed by rapid diffision of the mobil radioactive ions out of thestructure. The presence of concrete dissolution products such as calciumin the water in the pores will further accelerate the release ofradioactive ions from the zeolite into the environment.

In summary, it is demonstrated that waste form AX can serve as anextremely efficient barrier to isolate low and intermediate-level wasteconstituents from the environment. This material gives protection whichis as good as, and possibly better than, the protection againstextraction by water fixation. Not more than 0.16% of the initialactivity incorporated into AX can escape into the surroundings of thewater-form even under continuous exposure to water. On the other hand,even the best materials based on cement are shown to permit leakage ofthe bulk of the initial activity into the environment within a period ofa few years. They are, therefore, definitely unsuitable for use undercircumstances which allow for even a slight possibility of contactbetween the waste-form and an environment having water, wet soil, or ahumid atmosphere.

In conclusion, the chemical durability of the monolith should be greaterthan (and the leach rate therefore less than) about 10⁻⁵ g/cm⁻² /day at25° C. Preferably, the chemical durability of the monolith is greaterthan 10⁻⁶ g/cm² /day at 25° C.

EXAMPLE I

This example serves to illustrate a number of features of the invention.The ion exchange media is a glass based ion exchanger that is useful inremoving radioactive cations such as Cs-137 from solution. The ionexchange media by itself sinters at a temperature in excess of 1300° C.Such high temperatures are undesirable from two standpoints: (i) a largefraction of the radioactive Cs-137 will volatilize and will not betrapped in the resulting monolith; and (ii) the common alloys whichwould be preferred for cannister construction such as 304 stainlesssteel or Inconel cannot be used. In this example, the sinteringtemperature necessary to produce a monolith is lowered to 1050° C. bythe addition of a sintering aid, Additive A. The additive and ionexchange media are mixed dry and loaded in a 304 stainless steel column.Separation of the mixture during subsequent ion exchange operations isprevented by confining the mixture between two stainless steel frits.After the ion exchange process is complete, sintering is carried out insitu under water in a method suitable for applications where theradioactive loading on the column requires the shielding provided by alarge pool of water. The final waste form produced consists of a highlydurable glass monolith encased in a stainless steel cannister. Such awaste form would provide maximum protection against radioactive releaseduring shipment and burial.

Ion exchange Medium X was prepared as follows:

An alkali-borosilicate glass having the following nominal composition(on weight basis) was melted at 1400° C. in a platinum crucible:

SiO₂ : 60%

B₂ O₃ : 32%

Na₂ O: 3%

K₂ O: 5%

Cylindrical rods were drawn from the melt, crushed and sieved to collectgrains having particle sizes in the range of 350 to 710 micrometers, andheated for 2 hours at 550° C. This treatment resulted in phaseseparation and an interconnected microstructure. The resulting materialwas then treated with 3 N HCl at 95° C. whereby the alkali-borate richphase was leached leaving a porous glass consisting of approximately 95weight percent SiO₂ and 5 weight percent B₂ O₃ with trace amounts of Na₂O and K₂ O. After leaching, the residual HCl was removed by rinsing withdeionized water. The resulting porous glass powder was then placed in anaqueous solution of 3.5 molar NaNO₃ and 3.75 molar NH₄ OH at a pH ofapproximately 11.9 in order to replace protons on the hydroloyzed poresurfaces of the glass with sodium ions (Na⁺) and ammonium ions (NH₄ ⁺)by ion exchange. The Na⁺ and NH₄ ⁺ ions can in turn be replaced by othercations making Medium X an effective inorganic ion exchanger.

Twenty (20) cm³ of Medium X was vacuum dried overnight at roomtemperature. Additive A, a glass of composition given in Table II, wascrushed and sieved to produce grains having particle sizes of 210 to 355μm.

                  TABLE II                                                        ______________________________________                                        Composition of Additives (Weight Percent Oxide Basis)                                 Additive                                                              Oxide     A      B       C    D      E    F                                   ______________________________________                                        SiO.sub.2 42     81      46.0 40.0        72                                  B.sub.2 O.sub.3  13           9.5    38.8                                     PbO       49             45.32       61.2                                     Al.sub.2 O.sub.3  2                       2                                   Na.sub.2 O                                                                              2       4      2.5  12.9        14                                  K.sub.2 O 6              5.62 .1          1                                   MgO                                                                           Li.sub.2 O                                                                              1                                                                   CaO                           2.0         7                                   Rare Earth                    8.1                                             Oxides                                                                        Fe.sub.2 O.sub.3              11.1                                            ZnO                           5.0                                             TiO.sub.2                     3.0                                             MoO.sub.2                     2.4                                             ZrO.sub.2                     1.9                                             BaO                           0.6                                             P.sub.2 O.sub.5               0.5                                             Cs.sub.2 O                    1.1                                             SrO                           0.4                                             NiO                           0.6                                             Cr.sub.2 O.sub.3              0.4                                             TeO.sub.2                     0.3                                             CoO                           0.1                                             Other                    .56                                                  ______________________________________                                    

Four (4) cm³ of dried ion exchange Medium X was intimately mixed with 1cm³ of the powdered Additive A. The combined mixture was gravity fedthrough a funnel into a 0.50 inch OD×0.035 inch wall, type 304 stainlesssteel tube of 211/4 inch length having a stainless steel filter disc of20 μm porosity supported on a retainer ring at a height of 71/2 inchfrom the bottom. The tube was tapped gently to settle the mixture ontothe filter disc. A second filter disc was placed in the column on top ofthe mixture. A spring retainer was inserted on top of the upper filterdisc to prevent vertical movement of the glass mixture within thecolumn.

Referring now to FIG. 4, the loaded column 130 was inserted into thefurnace assembly 124 having heating elements and vacuum sealed by meansof two vacuum fittings 107, 109 at either end of the furnace. Additionalfittings 106, 110 at the ends of the stainless steel column were used tovacuum seal the column to the upper inlet and lower outlet fluid flowlines. The entire column-furnace assembly was immersed in the verticalposition in a tank 128 containing approximately 25 gallons of tap water.Vacuum pump 121 is used to evacuate the furnace completely.

With vent valve assembly 100 removed and valve 102 closed, a 1050 ml.volume of a solution containing 3000 ppm boron, 1000 ppm sodium, and 2ppm cesium was poured into reservoir 101. All concentrations weredetermined by Atomic Absorption/Emission Spectroscopy (AA/ES). Anadditional 50 ml. volume of a Cs-137 solution having an activity ofapproximately 10 μCi was added to the solution reservoir 101 and thevent valve assembly 100 reinstalled. With valves 100, 102, 103, 104,105, 111, 114, 115 and 126 closed, the reservoir 112 was filled with 100ml. of millipore water and the diaphragm vacuum pump 117 was engaged andadjusted to give a 14" (Hg) vacuum. The column was evacuated by openingvalves 105 and 103 and backwashed by slowly opening valve 11 andmaintaining this valve in the open position until the water streamentering trap 113 was free of bubbles.

An activity detector was mounted outside of the water tank and directedtoward the column 130 at the height corresponding to the approximateposition of the ion exchange media and additive mixture.

With the backwashing procedure completed, valves 111 and 105 wereclosed. An influent sample was acquired by opening vent valve 100 andslowly opening valve 102 to allow the influent from reservoir 101 tofill completely the clear plastic line between valves 102 and 103. Valve102 was then completely opened and valves 103 and 105 were opened toallow the influent from the reservoir 101 to be drawn into trap 113.When no further bubbles emerged from the line connecting valve 105 andthe trap 113 an additional 50 ml. volume of solution was drawn into trap113. Valve 105 was then closed and diaphragm pump 117 was shut off. Trap113 was removed from the system and emptied. A sample vial was placed atthe end of the line from valve 105 and valve 105 was opened allowing theinfluent sample to flow by the force of gravity directly from thereservoir 101 into the sample vial. After a 10 ml. sample was collectedin this manner, valve 105 was closed and the trap at 113 wasreinstalled. A total of six separate activity concentrationdeterminations were conducted on the aliquots taken from the influentsample. Each aliquot was evaporated to dryness and counted on a gas flowproportional counter which had been calibrated with a Cs-137 referencesource whose activity was known to ±10%. The determined average influentconcentration was 6480 ±580 (one σ) picocuries per ml.

The excess deionized water in the backwash reservoir 112 was discardedand an effluent receiver was installed in its place. Valve 126 wasopened. Valve 111 was opened and adjusted to allow a flow rate ofapproximately 1.5 ml./min of gravity driven influent from reservoir 101to flow through the column and into the effluent receiver 112. Valve 111was periodically adjusted to maintain a constant flow rate ofapproximately 1.5 ml./min during the course of the ion-exchange. Afterthe content of reservoir 101 had passed through the column, valve 111was closed and diaphram pump 117 was turned on. Valve 114 was openedthereby drawing the residual solution in the column and attached linesinto trap 113. When liquid flow into trap 113 terminated, valve 104 wasopened and solution entrapped in the lines between valve 104 and trap113 was similarly withdrawn. Valves 104 and 114 were then closed andvalve 105 was opened thereby withdrawing entrapped solution from theline between valve 105 and trap 113. After all liquid flow terminated,valves 105, 103, 102 and 100 were closed, and the effluent receivervalve 112 was removed. Three activity concentration determinations wereconducted on samples from the effluent receiver. For each determinationa 10 ml. sample of the effluent was evaporated and the residual materialwas counted in the manner used for the influent samples. The determinedaverage effluent concentration was 9.19±1.5 (σ) picocuries per ml.

The total volume of solution which flowed through the column was 1100ml. which corresponds to a total activity of Cs-137 of (6.840±0.580×10⁻³μCi/ml) (1100 ml)=7.128±0.638 μCi. The total effluent activity was(9.19×10⁻⁶ μ Ci/ml) (1100 ml)=0.101±0.0017 μCi. The averagedecontamination factor, DF, for the experiment is thus 705±131.

To prepare the column for heating valves 111 and 114 were opened andwhen all liquid flow stopped valves 111 and 114 closed and diaphram pump117 was shut off. With valves 115 and 126 open and all other valvesclosed and with column roughing vacuum pump 120 operating, the columnwas vacuum dried at ambient temperature for 12 hours following which thetemperature of the oven surrounding the column was programmed up to1050° C. over a 36 hour period, held at 1050° C. for 1 hour and thenallowed to return to ambient temperature at its normal cooling rate overa 12 hour period. During the heating period pump 121 was turned on toevacuate the furnace assembly. The oven-column assembly was then takenout of the water tank and the column removed from the oven.

Prior to heating, the activity detector outside the water tank gave anincreasing count rate with time, which corresponded to the accumulationof activity within the column. After heating and disassembly, 73% of thetotal Cs-137 activity was determined to be in the sintered glass and 27%on the tube walls. Visual inspection of the sintered mixture showed itto be a glassy grey, rigid foam of limited porosity.

A sample of this material having a geometric surface area ofapproximately 2.66 cm³ was washed twice with deionized water, each timefor a period of 10 minutes and was then placed in 100 ml. of de-ionizedwater in a Teflon vessel at 24° C. The leachant, (de-ionized water) wasthen completely removed and replaced with fresh de-ionized water atintervals of 1, 3, 6, 21, and 29 days. The matrix dissolution rate wascalculated based on determination of the silica level in the usedleachant by the reduced silicomolybdate spectrophotometric method. Thedissolution rate dropped by a factor of 12 from the first day to thesecond day and by another factor of 3.1 from the second day to the 25thday. On the 25th day, the dissolution rate was slightly below 3×10⁻⁷ gcm⁻² d⁻¹, a value which is three orders of magnitude better than thebest cements and comparable to that of the borosilicate glass used inEurope for high level waste disposal.

EXAMPLE II

This example illustrates the use of a number of other powderedadditives. The example is carried out in a fashion similar to Example Iwith variation in the type and amount of additive, sintering time andsintering temperature. Among the properties desirable in an additive arethe following: (i) it must lower the sintering temperature of the ionexchange media and additive mixture enough to prevent sublimation of theradioactive subtances trapped on the ion exchange media and/ordeterioration of the column cannister; (ii) it must lower the viscosityof the mixture enough to allow sintering in a reasonable amount of time;(iii) it must not interfere with the ion exchange capacity andefficiency of the media; and (iv) it must form a highly durable monolithwhen sintered with the ion exchange media. As this example shows, thechoice of additive will necessitate an optimization of the aboveproperties. No single additive possesses all of the above properties tothe highest degree.

The composition of Additives A-F is given in Table II. Some relevantphysical properties are given in Table III.

                  TABLE III                                                       ______________________________________                                                          Viscosity                                                                     (poise) at                                                                              Viscosity                                                                             Leach                                            Sintering  Sintering (poise) at                                                                            Rate                                      Additive                                                                             Temp (°C.)                                                                        Temp.     1050° C.                                                                       (g cm.sup.-2 d.sup.-1)                    ______________________________________                                        A      600        7 × 10.sup.5                                                                      1 × 10.sup.2                                                                    8 × 10.sup.-6                       B      1000       7 × 10.sup.5                                                                      3 × 10.sup.5                                                                    1 × 10.sup.-7                       C      726        7 × 10.sup.5                                                                      1 × 10.sup.3                                                                    5 × 10.sup.-5                       D                           1.3 × 10.sup.2                                                                  7 × 10.sup.-6                       E      590        1 × 10.sup.4                                                                      1       5 × 10.sup.-5                       F      850        3 × 10.sup.5                                                                      8 × 10.sup.4                                ______________________________________                                    

Each of the additives was individually crushed and sieved to producegrains having particle sizes of 210 to 355 micrometers. The powderedadditives were individually mixed with dried Medium X (prepared as inExample I) in the proportions given in Table I. The mixtures were thenindividually loaded into stainless steel columns and the ion exchangeand sintering steps carried out as in Example I with the exception thatthe influent solution contained no radioactive tracer Cs-137. Allconcentrations were determined by AA/ES. After sintering, each tube wascut open and the results of sintering observed. The sinteringperformance was rated on a scale of 0 to 3 as given in Table I.

Additives containing large percentages of the low melting oxides oflead, bismuth, phosphorous, boron, alkaline earth metals or alkalimetals are effective at lowering the sintering temperature, whereasadditives containing large percentages of the high melting, refractoryoxides of silica, aluminum and zirconium should be ineffective. Theadditives contained in this example confirm the above expectations.Additives B and F contain a large percentage of SiO₂ as well as some Al₂O₃. Both produce no sintering at or below 1050° C. Additive E iscomposed only of B₂ O₃ and PbO. It produces excellent sintering at 1000°C. Additive A contains a relatively low percentage of SiO₂, a highpercentage of PbO and alkali oxides (Na₂ O, K₂ O, Li₂ O) and producesgood sintering at 1050° C. The composition of Additive C is similar tothat of A, but is somewhat lower in PbO and the alkali oxides, andhigher in SiO₂. It produces medium sintering at 1050° C. Additive D hasthe lowest percentage of SiO₂ of the additives listed, and a relativelyhigh percentage of B₂ O₃ and alkali oxides (Na₂) and some alkaline earthoxides (CaO, BaO). It too prduces good sintering at 1050° C. In terms ofability to reduce the sintering temperature, the additives of Table IImust be rated E>A=D>C>F>B. This order parallels that produced byordering the pure additives by their sintering temperatures as given inTable III (lowest E, A, C, F, B highest) or by their viscosities at1050° C. (lowest E, A˜D, C, F, B, highest).

None of the additives tested in this example appeared to interfere withthe ion exchange properties of the ion exchange media except bydilution. Additives such as A, D and E which can give good sintering atlow volume percents are to be preferred over others that produce goodsintering only at large volume percents. Undoubtedly, even Additive B,if present at greater than 99 volume % would produce good sintering withMedia X at 1050° C. Unfortunately, the ion exchange capacity of such amixture would be almost nil.

If the additive is present during the ion exchange process, as in thisexample, the calculations given in the description of the invention showthat its dissolution rate should be less than 10⁻⁴ g cm⁻² d⁻¹ andpreferably less than 10⁻⁵ g cm⁻¹ d⁻¹. Ranking the additives tested interms of dissolution rate we have: (lowest, B, A˜D, E˜C highest) makingB the preferred additive in terms of durability. Additives B, A and Dall have dissolution rates below the preferred value of 10⁻⁵ g cm⁻² d⁻¹.

Available information on dissolution rates vs chemical compositionindicates that SiO₂, Al₂ O₃ and ZrO₂ increase durability (decreasedissolution rates) while alkali metal oxides and B₂ O₃ decreasedurability. It is thus expected that Additives B and F should have verygood durability and impart good durability while Additive E should havepoor durability and impart poor durability to the final monolith. theexpectations are born out by Table II and by observations of the finalmonolith when tested as in Example I.

The final choice of powdered additive depends very much on the maximumsintering temperature which in turn depends on the sublimationtemperature of radioactive species trapped on the ion exchange media andthe maximum process temperature the cannister material can withstand.Once this temperature is determined, all additives that produce goodsintering at that temperature should be examined, and the one withhighest durability chosen. The following illustrations using the sixadditives in Table I may prove useful:

(A) If the maximum process temperature is 1250° C., all additives willgive good sintering at this temperature, and Additive B will be chosenif 50 volume % additive is acceptable since it has the highestdurability;

(B) If the maximum process temperature is 1050° C., Additives A, D, andE provide good sintering, and Additive A or D will be chosen since theyhave the highest durability; and

(C) If the maximum process temperature is 950° C., Additive E willproduce the best sintering, and Additive E should be chosen.

EXAMPLE III

This example serves to illustrate the use of two powdered additives, Aand E, with an inorganic ion exchange media Y, ionsiv IE95, presentlyused in nuclear waste water treatment. Ionsiv IE95 is a zeolite (sodiumalumino silicate) produced by the Linde Division of the Union CarbideCorp., that is particularly effective for removing Cs from solution. Byitself, it sinters at temperatures in excess of 1300° C. Therefore, formost applications, especially those involving entrapment of radioactiveCs, a sintering aid is required.

Additives A and E were individually crushed and sieved to produce grainshaving particle sizes of 210 to 355 micrometers. The powdered additiveswere individually mixed with dried Media Y (20×50 size as specified bythe manufacturer). The final mixtures contained 30% additive in eachcase. The mixtures were then individually loaded into stainless steelcolumns and the ion exchange and sintering steps carried out as inExample I using the influent solution of Example II. No Cs was detectedin the effluent during the course of the ion exchange process. Bothmixtures were sintered at 1050° C. After sintering and cooling each tubewas cut open and the sintering performance of each mixture rated as inExample II. In both cases the mixtures produced well sintered glasseswith very little porosity and were given a performance rating of three.In this case, since both additives gave good sintering at 1050° C.,Additive A is preferred because of its higher durability. This examplecould be repeated using radioactive Cs with similar results.

EXAMPLE IV

This example demonstrates the use of an additive, A, with an inorganicanion exchanger, Media Z which is prepared as follows: Rods of thealkali-borosilicate glass prepared as in Example I were cut into 3"sections, then they were leached as in Example I for 24 hours and werewashed. Six of these porous glass rods were immersed in an 11.7%Zr(NO₃)₄.5H₂ O aqueous solution at room temperature for 17 hours thusallowing the Zr(NO₃)₄ to diffuse inside the pores of the glass. Thestuffed rods were then transferred to an oven at 100° C. for 11/2 hoursto evoke precipitation of the Zr salt by evaporation of the water. Therods were heated to 200° C. under vacuum to decompose the nitrate withinthe glass pores into zirconium oxide which hydrates in the presence ofwater to impart anionic exchange capability. It is believed that thehydrated zirconium atoms bonded to each other in the form of crystalsand that some of the zirconium atoms are bonded to silicon of the glassrod through divalent oxygen linkages. The rods were then crushed andsieved to collect the 355-710 μm fraction. This material is ion exchangeMedia Z.

Ion exchange and sintering tests were carried out on a mixture of 20%Additive A and 80% Media Z according to the procedures in Example I withthe following changes: (1) the influent solution was 1.0 ppm CrO₄ ⁻² ion(as Na₂ CrO₄) and 1.0 ppm MoO₄ ⁻² ion (as Na₂ MoO₄) in de-ionized water;(2) since no radioactive tracers were used, all analyticaldeterminations were made using D.C. plasma emission spectroscopy; (3)analysis of the effluent solutions showed that over 90% of the CrO₄ ⁻²and MoO₄ ⁻² anions present in the influent had been absorbed on Media Z.This example could be repeated using radioactive Cr and Mo with similarresults.

EXAMPLE V

Successful sintering requires that an intimate homogeneous mix ofadditive and ion exchange media be effected prior to sintering. Caremust also be taken that no separation of the mixture occur during suchsteps as backwashing and ion exchange. The previous examplesdemonstrated that dry mixing of additive and media produces the requrieduniform, homogeneous mixture and that rigid confinement of the mixturebetween two porous discs prevents separation during backwashing and ionexchange. This example illustrates the problem of mixture separationduring backflow operations and describes another technique to preventit. This example also illustrates techniques that can be used to mixadditive and ion exchange media in slurry form.

The following initial tests were carried out: (i) Additive D having thecomposition set forth in Table II and a bulk density of 3.0 grams/cm³was ground and sieved to obtain particles within the size range of 350to 550 micrometers. Five cm³ of these particles was mixed in a slurrywith 5 cm³ of ion exchange Media X prepared as in Example I which had abulk density of 1.7 grams/cm³. The mixture was then placed in atransparent, plastic column having a cross sectional area of 0.89 cm²and backwashed by passing a stream of deionized water from the bottom tothe top of the column at the rate of 43 cm³ /min. for 5 minutes. Uponcompletion of the backwash step, it was observed that the additive(black in color) remained at the bottom of the column while the ionexchange medium (white) had risen to the top of the column.

(ii) Five cm³ of an ion exchange Media X prepared as in Example I wassoaked in a solution of cobalt ion Co⁺² (100 ppm Co(NO₃)₂). Theresulting blue media, in slurry form, was mixed with 5 cm³ of Additive Ahaving the composition set forth in Table II which had a bulk density of3.84 grams/cm.³ and particle size ranging from 177 to 250 micrometers.Once again the additive and the ion exchange medium separated with theadditive remaining in the bottom of the column while the ion exchangemedium had risen to the top.

(iii) Five cm³ of ion exchange Media X made as in Example I and having awet bulk density of 1.7 grams/cm³ and 5 cm³ of Additive D having thecompositin defined in Table II, a bulk density of 3.0 grams/cm³ andparticle sizes within the range of 125 to 355 micrometers was placedinto a column having a cross sectional area of 0.89 cm². As originallyplaced in the column, ion exchange media and additive were not uniformlymixed. The column was then backwashed with de-ionized water at the rateof 40 cm³ /min. for 5 minutes. A slight amount of Additive D wasobserved at the top of the column at the conclusion of the backwashoperation, but otherwise the ion exchange media and the additive wereuniformly mixed in the column.

(iv) Five cm³ of ion exchange Media X prepared as in Example I wassoaked in a solution of cobalt ion (Co⁺² as in (ii)). The resulting blueion exchange media, in slurry form, was mixed with 5 cm³ of Additive Ahaving the composition defined in Table II, a bulk density of 3.84grams/cm³, and particle sizes within the range of 90-125 microns, andplaced in a column as above. After 5 minutes of backwash at the rate of40 cm³ /min., the mixture was uniform with little or no separation ofmedia and additive.

Each of the four mixtures given in Table IV was prepared. Note thatthese mixtures parallel the four tested above in the transparent plasticcolumn.

                  TABLE IV                                                        ______________________________________                                        Mix-          Media            Additive                                                                              Volume %                               ture  Media   Size (μm)                                                                            Additive                                                                             Size (μm)                                                                          Additive                               ______________________________________                                        1     X       355-710   D      350-550 50                                     2     X       355-710   A      177-250 50                                     3     X       355-710   D      125-355 50                                     4     X       355-170   A       90-125 50                                     ______________________________________                                    

Each mixture was loaded in a stainless steel tube as in Example I. Thistime, however, the second filter disc was not placed in the column ontop of the mixture, thus allowing the mixture to expand during backwash.Each column was backwashed at a rate of 40 cm³ /min for 5 minutes priorto ion exchange. Ion exchange and sintering was then carried out as inExample I with the exception that no Cs-137 tracer was used, allconcentrations was determined by AA/ES. After cooling each tube was cutopen and the results of sintering observed. Mixtures 1 and 2 hadevidently separated during the backwashing process. The additive at thebottom of the mixture had sintered very well. However, the ion exchangemedia had not sintered at all. Mixtures 3 and 4 evidently had notseparated during backwashing and were well sintered (performance index3).

This example illustrates that careful selection of the media andadditive particle size according to the guideline given in thedescription of the invention vis.

    ρ.sub.A S.sub.A.sup.1/2 =ρ.sub.M S.sub.M.sup.1/2

where ρ is the bulk density, S is the particle size and A and M standfor additive and ion exchange media respectively, can prevent separationduring counter-flow operations such as backwashing. The example alsoshows that given the proper choice of particle sizes, counter-flowoperations can actually promote uniform mixing.

EXAMPLE VI

This example illustrates another method of preventing separation of theadditive and media. The mixture separation is prevented by modifying thebackwash procedure in one of two ways (i) Four cm³ of the ion exchangeMedia X which had previously been soaked in a cobalt (Co⁺²) solution asin Example V and then dried, and 1 cm³ of Additive A having particlesize in the range of 177 to 250 micrometers was mixed and then loadeddry into a glass column having a cross sectional area of 0.95 cm². Noseparation of additive and media was observed. In order to remove airpockets from the mixture which would cause "channeling" during forwardflow operation, the mixture was gently backwashed at a rate of 1 ml/min.with water. During backwash the mixture did not expand and no separationof the mixture was observed. (ii) A mixture of ion exchange Media X andAdditive A is prepared and loaded in a column as in (i) above. Airpockets in the mixture are removed by evacuating the column and thengently sucking water into the column. No separation of ion exchangemedia and additive occurs.

A mixture of 80 vol % Media X and 20% Additive A (particle size 177 to250 μm) was loaded into a stainless steel tube as in Example V. Thebackwash rate was changed to 1 ml.min. as in (i) above. After thebackwash the example proceeded as Example V through sintering.Inspection of the mixture after sintering showed a well sintered productwith no evidence of mixture separation. Examination of the effluentconcentrations gave no evidence of channeling during the ion exchangeprocess since capacity and DF figures were similar to those in ExampleI.

The procedure in the above paragraph was repeated with the backwashmodified as in (ii). The results were identical to those obtained above.Thus, slowing down or eliminating back flow operations can preventmixture separation.

EXAMPLE VII

This example demonstrates the introduction of the additive in liquidform after the ion exchange process has taken place. This method ofintroduction of the additive avoids the dilution of the ion exchangemedia by additive during the ion exchange process. It also eliminatesthe possibility of interference by the additive with the ion exchangecapacity or efficiency of the ion exchange media. Finally, it leaves theadditive behind as a uniform coating on the media or as a powderuniformly mixed with the ion exchange media.

Liquid additives are produced by dissolving a suitable sintering aid ina solvent which can subsquently be evaporated cleanly leaving thesintering aid behind. After the ion exchange process is completed theion exchange media is dried and enough liquid additive introduced tocover the ion exchange media. The solvent is then evaporated leaving theaid behind. This process can be repeated several times to increase theamount of aid deposited in the ion exchange media.

A number of liquid additives are given in Table V.

                  TABLE V                                                         ______________________________________                                        Liquid                 Liquid Additive                                        Additive                                                                             Solvent         Composition/100 ml. soln                               ______________________________________                                        H      1.3 molar NH.sub.4 OH                                                                         18 g H.sub.3 BO.sub.3                                         (aqueous)                                                              I      Methanol        18.2 g H.sub.3 BO.sub.3                                J      5.8 molar NH.sub.4 OH                                                                         20.8 g H.sub.3 BO.sub.3 + 10.4 g                              (aqueous)       Zn(OH).sub.2                                           K      8.2 molar NH.sub.4 OH                                                                         20 g H.sub.3 BO.sub.3 + 26 g                                  (aqueous)       CoCl.sub.2.6H.sub.2 O                                  L      Water           41 g Pb(C.sub.2 H.sub.3 O.sub.2).sub.2.3H.sub.2 O      M      Hot water (70° C.)                                                                     69 g Pb(C.sub.2 H.sub.3 O.sub.2).sub.2.3H.sub.2 O      N      Methanol        12.1 g H.sub.3 BO.sub.3 + 38 g                                                Mg(NO.sub.3).6H.sub.2 O                                O      Water           52 g Pb(NO.sub.3).sub.2                                P      17 molar acetic acid                                                                          86.4 g Bi(NO.sub.3).sub.3.5H.sub.2 O                   ______________________________________                                    

Ion exchange Media X is prepared as in Example I. It is then placed inan ion exchange column consisting of a 304 stainless steel tube similarto that used in Example I but with a small dead volume below the bottomfrit. Five cm³ of ion exchange media is placed in the tube on top of thebottom frit and a second frit is forced down on top and held in place bya second clip. The column is then placed in the tubing and furnacesystem as in Example I. The column is then backwashed and influentsimilar to Example I passed through the column as in Example I. Afterthe flow of influent stopped the system was sampled and vacuum dried for12 hours as in Example I. All valves were then closed. Reservoir 112 wasthen filled with liquid additive H and pump 117 turned on. Valves 111,126 and 105 were opened and enough of the liquid Additive H was suckedup into the column to just cover the top frit. Valves 111, 126, and 105were then closed and the system pumped on by opening valves 104 and 115for 24 hours, evaporating the solvent in liquid Additive H. The furnacewas evacuated and the temperature then raised at 1° C./minute to 1050°C. and held there for 1 hour, thus sintering the glass to form a clearhomogeneous foam glass that has trapped the radioactive Cs.

A sample of the glass was tested for durability as in Example I withleachant (deionized water) replacement taking place at 2, 5, 16, 23, 30and 37 days. The matrix (silica) dissolution rate after 37 days was5.3×⁻⁷ g.cm⁻².d⁻¹. This dissolution rate compares favorably with that ofthe borosilicate glass presently used in Europe for high level wastedisposal.

EXAMPLE VIII

In this example, eight liquid additives (I-P of Table VI) are tested.The example is identical to Example VII except that Additives I-P wereindividually used in place of Additive H and that the influent containedno Cs-137 tracer. All Cs concentrations were determined by Aa/Es.

The results of the sintering process are given in Table VII.

                  TABLE VII                                                       ______________________________________                                        Effectiveness of Liquid Additives                                             Liquid                                                                        Additive                                                                             Result of Sintering                                                    ______________________________________                                        H      Clear homogeneous foam glass produced.                                 I      Black polycrystalline foam produced. Charcoal                                 like consistancy.                                                      J      Clear homogeneous foam glass produced.                                 K      Grainy foam produced. Blue (cobalt) not uniformly                             distributed.                                                           L      Black glassy material produced which powders                                  easily.                                                                M      Black glassy material produced. Powders easily                                with small beads of metal present.                                     N      Black crystalline material produced which powders                             easily.                                                                O      White grainy foam glass produced. Not                                         homogeneous which powders easily.                                      P      Grainy glass produced which powders easily.                            ______________________________________                                    

The liquid additives using methanol for a solvent (I and N) gave poorend products that were black and powdered easily. The color andmechanical weakness are probably caused by carbon from methanol that wasnot removed prior to sintering. This problem may possibly be overcome bymore careful evaporation of the solvent (i.e., longer times and/orslower heating rates). It may also be ion exchange media dependant andmay not be a problem with other ion exchange media. A similar poorend-product was produced when Pb(C₂ H₃ O₂)₂ 3H₂ O was used as theAdditive (L, M), and probably results from carbonization of the acetate(C₂ H₃ O₂). This problem may also possibly be overcome by altering theevaporation of heating process. The additives containing H₃ BO₃ inaqueous NH₄ OH (H, J and K) produced good end products which were foamswith non interconnected pores. As noted with the solid additives, B₂ O₃/H₃ BO₃ is an excellent sintering aid. However, it can not be used aloneas a solid additive because of its poor chemical durability. The Pb andBi nitrate additives (O, P) seem to produce mechanically weak foamswhich are poor end-products. Samples produced using liquid additives J,K and L were tested for durability as in Example VII. The matrix(silica) dissolution rates after 37 days are listed in Table VIII.

                  TABLE VIII                                                      ______________________________________                                        Liquid Additive                                                                          Matrix Dissolution Rate (μg cm.sup.-2 day.sup.-1)               ______________________________________                                        J          5.37                                                               K          1.30                                                               L          2.73                                                               ______________________________________                                    

As can be seen, the durability of each product is quite high.

As will be readily understood by those of ordinary skill in the art,minor modifications may be made in the process and apparatus describedabove without in any way departing from the spirit and scope of theinvention. Accordingly, it is understood that the invention will not belimited to the exact details disclosed hereinabove, but will be definedin accordance with the appended claims.

What is claimed is:
 1. A method of removing radioactive ions from aliquid containing such ions with an ion exchange medium and sinteringthe ion exchange medium to form a monolith, comprising:(1) providing anion exchange medium for radioactive ions, said ion exchange medium beingselected from the group consisting of:(i) porous silicate glass orporous silica gel, (ii) natural or syntheitc clays, (iii) hydrated metaloxides, (iv) the alkali salts of said hydrated metal oxides, and (v)mixtures of (i)-(iv); (2) containing said ion exchange medium in acanister capable of withstanding the sintering temperature of said ionexchange composition; (3) contacting said ion exchange medium containedin said canister with said liquid to exchange cations or anions of saidion exchange medium with said radioactive ions; (4) mixing said ionexchange medium contained in said canister with an additive which is asintering aid for said ion exchange medium and permits said compositionto be sintered at a temperature below about 1200 degrees C., saidadditive being mixed with said ion exchange medium either before orafter step (3); and (5) sintering said ion exchange medium contained insaid canister to form a monolith.
 2. A method according to claim 1wherein the additive is dissolved in a liquid solvent and said additiveis impreganted on said ion exchange medium by the steps of:contactingsaid additive containing liquid solvent with the ion exchange mediumcontained in said canister; and withdrawing said liquid solvent fromsaid ion exchange medium leaving behind said additive as a coating onthe ion exchange medium contained in said canister.
 3. A methodaccording to claim 1 wherein said additive is a powder and ismechanically mixed with said ion exchange medium contained in saidcanister.
 4. A method according to claim 1 wherein said additive isselected from the group consisting of alkali metal oxides, alkalineearth metal oxides, SiO₂, B₂ O₃, PbO, P₂ O₅, Bi₂ O₃, Nd₂ O₃, Fe₂ O₃,ZnO, TiO₂, MoO₃, ZrO₂, CoO or mixtures thereof.
 5. The method of claim 1wherein the ion exchange medium contained in said canister is mixed withthe additive to ensure that a homogeneous mixture is present, andwherein said homogeneous composition is heated to its sinteringtemperature.
 6. A method according to claim 5 wherein said heating iscarried out under vacuum.
 7. A method according to claim 5 wherein saidmixing step further includes the introduction of carbon dioxide into theion exchange mixture contained in said cannister.
 8. A method of claim 1wherein said cannister is made of stainless steel.
 9. A method of claim1 wherein asid monolith has a chemical durability of greater than 10⁻⁵g/cm² /day at 25° C.